Influence of steam flow rate on oxidation kinetics of silicon carbide at 1400-1600 C

Pham V. H.; 倉田 正輝; 永江 勇二; 石橋 良*; 佐々木 政名*

Corrosion Science, 255, p.113098_1 - 113098_9, 2025/10

https://doi.org/10.1016/j.corsci.2025.113098

Being expected as materials for accident tolerant fuel cladding tube, oxidation behavior and kinetics of silicon carbide (SiC) under extreme conditions like severe accidents must be elucidated. In this study, oxidation tests of SiC at 1400-1600 C for 1-5 h, at atmospheric pressure, under two different flow rates of HO/Ar gas mixture have been conducted to investigate the influence of steam flow rate on the formation of SiO scale and its subsequent volatilization. The oxidation tests were conducted via a newly developed test facility using laser as a heat source. Oxidation kinetics of SiC was evaluated via mass change of samples before and after the oxidation tests. Parabolic oxidation rate representative for SiO formation and linear volatilization rate reflecting its volatilization were calculated, based on these mass changes. The Arrhenius dependence of the parabolic oxidation and linear volatilization rate constants were then plotted. Results of this study indicated that SiC exhibits excellent performance under the conditions investigated. Steam flow rate has a significant influence on volatilization of SiO but has minor effects over its formation. Oxidation of SiC in steam at high temperature may follow mass gain or mass loss regime, depending on the steam flow rate. Two oxidation patterns were suggested and discussed. In the first oxidation pattern, the SiO formation is dominated over its volatilization. The second oxidation pattern (steady stage) is reached when the SiO formation rate is equivalent to its volatilization rate. Time to reach this steady stage was defined, based on the parabolic oxidation rate and linear volatilization rate.

燃料デブリ分析のための超微量分析技術の開発(委託研究); 令和5年度英知を結集した原子力科学技術・人材育成推進事業

廃炉環境国際共同研究センター; 東北大学*

JAEA-Review 2024-064, 118 Pages, 2025/06

JAEA-Review-2024-064.pdf:6.73MB

日本原子力研究開発機構(JAEA)廃炉環境国際共同研究センター(CLADS)では、令和5年度英知を結集した原子力科学技術・人材育成推進事業(以下、「本事業」という。)を実施している。本事業は、東京電力ホールディングス株式会社福島第一原子力発電所(1F)の廃炉等をはじめとした原子力分野の課題解決に貢献するため、国内外の英知を結集し、様々な分野の知見や経験を、従前の機関や分野の壁を越えて緊密に融合・連携させた基礎的・基盤的研究及び人材育成を推進することを目的としている。平成30年度の新規採択課題から実施主体を文部科学省からJAEAに移行することで、JAEAとアカデミアとの連携を強化し、廃炉に資する中長期的な研究開発・人材育成をより安定的かつ継続的に実施する体制を構築した。本研究は、令和元年度に採択された研究課題のうち、「燃料デブリ分析のための超微量分析技術の開発」の令和元年度から令和5年度分の研究成果について取りまとめたものである。本研究は、燃料デブリの取り扱い、臨界管理、保管管理等に必要な性状把握において、キーとなるアクチノイド核種の化学分析を中心に、最適な試料前処理・分離・分析プロセスを開発し、将来計画されている燃料デブリ分析の効率化・合理化を図るとともに、一連の研究業務における人材育成を通し、1F廃炉推進に資することを目的とする。特に、近年分析化学分野、放射化学分野で成果を上げつつある誘導結合プラズマ質量分析法(ICP-MS/MS)を原子力分野に応用することにより、測定核種を単離するための前処理をせずに高精度で分析できる手法を開発し、分離前処理を省力化し、迅速な分析工程を確立するとともに大学、企業を含めた体制が構築された。

Development of a method for the determination of spontaneous fission nuclides in irradiated fuel and applicability to Pu quantification in fuel debris by dual time neutron measurements

長谷 竹晃; 小菅 義広*; 相楽 洋*; 中岫 翔; 能見 貴佳; 奥村 啓介

Progress in Nuclear Science and Technology (Internet), 7, p.41 - 46, 2025/05

https://doi.org/10.15669/pnst.7.41

This paper provides an overview of plutonium quantification in irradiated fuel including fuel debris at the Fukushima Daiichi Nuclear Power Plants, named Dual Time Measurement (DTM) method. Spontaneous fission nuclides in irradiated fuel decrease exponentially with the passage of time according to the mainly half-life of Cm-244 (half-life of about 18.11 years). By measuring neutrons two times with long time intervals, Pu-240 effective mass (half-life of about 6,500 years) and Cm-244 mass can be quantified. Pu mass can be quantified by utilizing the correlation between ratio of Cm-244/ Pu-240 effective mass and Pu/ Pu-240 effective mass. The applicability of DTM method was evaluated numerically. The results show that long time interval was required to reduce the random errors. In the case that the interval between the first and second measurements is 32 years, Pu-240 effective mass and Pu can be quantified with uncertainties of 10-50% depending on the presence of water in storage canister and the burnup condition of irradiated fuel including the mixture of several burnup compositions in fuel debris.

Uncertainty quantification of Np, Am, and Am reaction rates in highly enriched uranium fuel cores at Kyoto University Critical Assembly

Pyeon, C. H.*; 大泉 昭人; 方野 量太; 福島 昌宏

Nuclear Science and Engineering, 199(3), p.429 - 444, 2025/03

https://doi.org/10.1080/00295639.2024.2380624

京都大学KUCAの高濃縮ウラン燃料による複数の炉心で取得したNp、Am、Amの核分裂率とNp捕獲反応率に関する積分実験データに対して、ENDF/B-VIII.0及びJENDL-5とSerpent2コードを用いた実験解析を実施した。Np/U、Am/U、Am/Uの核分裂率比の実験値と解析値の比較では、ENDF/B-VIII.0及びJENDL-5のいずれにおいて、それぞれ約5、15、および10%の精度が示された。Np/Auの捕獲反応率比については、熱中性子スペクトルの炉心において高い精度が得られることを明らかにした。また、Np/U、Am/U、Am/Uの核データに起因した不確かさは、概ね4%以内であることを示した。

Characterization of neutrons emitted by an expected small amount of fuel debris in a trial retrieval from Fukushima Daiichi Nuclear Power Station

松村 太伊知; 奥村 啓介; 坂本 雅洋; 寺島 顕一; Riyana E. S.; 近藤 千博*

Nuclear Engineering and Design, 432, p.113791_1 - 113791_9, 2025/02

https://doi.org/10.1016/j.nucengdes.2024.113791

Retrieving objects with a small amount of fuel debris, such as a few grams, will begin soon at the Fukushima Daiichi Nuclear Power Station (1F) at the start of decommissioning. Objects retrieved from the primary containment vessel are not necessarily fuel debris; fuel debris is an object from which neutrons are emitted because it contains nuclear-fuel material. However, the characteristics of the neutrons emitted by fuel debris are unknown. Fuel debris was categorized into five types according to the elapsed time from the accident, burnup, and fuel type (UO or mixed oxide). The number and energy spectra of (, ) and spontaneous fission neutrons emitted from 1 g of each fuel debris type were estimated using the SOURCES 4C code to obtain the neutron characteristics. The results showed that the average neutron energy is approximately 2.1 MeV, regardless of the type of fuel debris. However, the intensities of neutrons emitted from the fuel debris in 1F Units 2 and 3 varied by four orders of magnitude according to the fuel debris type.

Development of a theoretical scaling factor method for the inventory estimation of difficult-to-measure nuclide Cs-135 in fuel debris and radioactive wastes

坂本 雅洋; 奥村 啓介; 神野 郁夫; 松村 太伊知; 寺島 顕一; Riyana E. S.; 金子 純一*; 溝上 暢人*; 溝上 伸也*

Journal of Nuclear Science and Technology, 10 Pages, 2025/00

https://doi.org/10.1080/00223131.2025.2478926

In this paper, we propose a new nuclide inventory estimation method based on computational methods, called a "theoretical scaling factor method" for difficult-to-measure (DTM) nuclides in fuel debris and radioactive wastes. The theoretical scaling factor method provides a method similar to a conventional scaling factor method. The theoretical scaling factor method, however, does not require performing many measurements to obtain correlations between a key nuclide which is easy-to-measure and a DTM nuclide. Instead of actual analytical measurements, the results of theoretical calculations are used. A correlation equation between the key nuclide and the DTM nuclide is created based on the results of theoretical calculations, and the DTM nuclide is deterministically estimated using the measurement value of the key nuclide only. In this paper, we selected Cs-135 as the DTM nuclide and Cs-137 as the key nuclide. Cs-135 has a long half-life of 2.310 years and is one of the important fission products in the safety evaluation for the geological disposal of high-level radioactive waste, because it dissolves and migrates in groundwater easily. We confirmed the validity of the proposed method using measured data of Cs-137 and Cs-135 on radioactive wastes from the Fukushima Daiichi Nuclear Power Station (1F) accident obtained by many researchers. It can be used as a rational and efficient technology to reduce the analysis costs of various types of fuel debris and radioactive waste present at 1F.

Effect of sample temperature and laser ablation angle on optical emission and acoustic signals from laser-induced Zirconium plasma

Batsaikhan, M.; 大場 弘則; 狩野 貴宏; 赤岡 克昭; 若井田 育夫

Optics Express (Internet), 32(24), p.42626 - 42638, 2024/11

https://doi.org/10.1364/OE.538774

A fiber-coupled, acoustic-wave-assisted microchip laser-induced breakdown spectroscopy system (AW-mLIBS) was developed to analyze the elemental composition and surface imaging. In this study, we measured the dependence of sample temperature and laser ablation angle on the laser-induced plasma-optical emission and LIP-acoustic signal. The intensity of the laser-induced plasma-optical emission and ablated mass at three different temperatures and eight different laser ablation angles were estimated using a zirconium sample. Simultaneously, we investigated the laser-induced plasma-acoustic signal amplitude, propagation speed, and shape by synchronizing the AW-mLIBS system with a high-speed camera. The results revealed that the laser-induced plasma-optical emission increases with increasing temperature and is unaffected by laser ablation angle up to 40 degree because the amount of the ablated mass was similar and the plasma. Additionally, no considerable variation in plasma temperature obtained using the Boltzmann method over the sample temperature. However, the propagation speed of the laser-induced plasma-acoustic signal differs with temperature but has marginal angular dependence because the laser-induced plasma-acoustic signal propagates as semispherical. Furthermore, small and no considerable changes observed in the laser-induced plasma-acoustic signal amplitude up to 100 degree Celsius, and the laser ablation angle showed a similar tendency to that of the laser-induced plasma-optical emission.

高速炉燃料用SUS316相当鋼の高温強度及び照射特性評価

宮澤 健; 上羽 智之; 矢野 康英; 丹野 敬嗣; 大塚 智史; 鬼澤 高志; 安藤 勝訓; 皆藤 威二

JAEA-Technology 2024-009, 140 Pages, 2024/10

JAEA-Technology-2024-009.pdf:8.03MB

SUS316相当鋼を用いた高速炉燃料設計の高信頼性化に向けて、SUS316相当鋼被覆管及びラッパ管の高温強度及び照射データを材料学的及び統計学的な観点で評価・解析することで、高温強度及び高照射量までの照射特性に係る設計用強度式を導出した。異常な過渡変化の上限温度を超える900CまでのSUS316相当鋼被覆管及びラッパ管(非照射材)の高温引張試験データ及び高温クリープ試験データを拡充し、0.2%耐力、引張強さ、クリープ破断強度の最適近似式と下限式並びに熱クリープひずみの最適近似式と上下限式を導出した。また、高速実験炉「常陽」、仏国・高速原型炉Phenix及び米国・FFTFで高照射量まで中性子照射したSUS316相当鋼被覆管及びラッパ管の照射後引張試験データ及びSUS316相当鋼被覆管の炉内クリープ破断試験データを解析することで、炉内Na中照射による引張強度及びクリープ強度の低下を表す強度補正係数を導出した。導出した式を実測値と比較することで、その妥当性を確認した。

An Introduction to spent nuclear fuel decay heat for light water reactors; A Review from the NEA WPNCS

Rochman, D.*; 湊 太志; 渡邉 友章; 他52名*

EPJ Nuclear Sciences & Technologies (Internet), 10, p.9_1 - 9_83, 2024/10

https://doi.org/10.1051/epjn/2024010

This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments.

Initial verification of Cyclus and NMB fuel cycle simulators

Bachmann, A. M.*; Richards, S.*; Feng, B.*; 西原 健司; 阿部 拓海

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

この研究は、燃料サイクルシミュレーションを活用するための初期段階として、コード検証を行ったものである。CyclusとNMBは、核燃料サイクルオプションの解析を提供するオープンソースの燃料サイクルシミュレータであり、アルゴンヌ国立研究所と日本原子力研究開発機構(JAEA)によって、燃料サイクルベンチマークに関する複数年の共同研究のためにそれぞれ選ばれた。両者とも比較的新しく、厳密なコード間比較を行った後に改良することが可能である。これらのシミュレータの初期検証は、ワンススルー及びマルチリサイクル燃料サイクルの一連の仮想シナリオを用いて実施した。その結果、2つのシミュレータのシナリオ定義とモデル化手法の違いが、物質インベントリ、マスフロー、及び燃料サイクル評価における他の重要な測定基準における結果の違いにどのようにつながるかが明らかになった。